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The compositional effects of ternary Zr-Cu-X (X: Al, Fe) amorphous filler alloys on galvanic corrosion susceptibility in high-temperature pressurized water were investigated for Zircaloy-4 brazed joints. Through an Al-induced microgalvanic reaction that deteriorated the overall nobility of the joint, application of the Zr-Cu-Al filler alloy caused galvanic coupling to develop readily between the Al-bearing joint and the Al-free base metal, finally leading to massive localized corrosion of the joint. Contrastingly, joints prepared with a Zr-Cu-Fe filler alloy showed excellent corrosion resistance comparable to that of the Zircaloy-4 base metal, since the Cu and Fe elements forming fine intermetallic particles with Zr did not influence the electrochemical stability of the resultant joints. The present results demonstrate that Fe is a more suitable alloying element than Al for brazing filler alloys subjected to high-temperature corrosive environments. [ABSTRACT FROM AUTHOR]

The authors previously conducted thermogravimetric analyses on Zircaloy-2 in air. By using the thermogravimetric data, an oxidation model was constructed in this study so that it can be applied for the modeling of cladding degradation in spent fuel pool (SFP) severe accident condition. For its validation, oxidation tests of long cladding tube were conducted, and computational fluid dynamics analyses using the constructed oxidation model were proceeded to simulate the experiments. In the oxidation tests, high temperature thermal gradient along the cladding axis was applied and air flow rates in testing chamber were controlled to simulate hypothetical SFP accidents. The analytical outputs successfully reproduced the growth of oxide film and porous oxide layer on the claddings in oxidation tests, and validity of the oxidation model was proved. Influence of air flow rate for the oxidation behavior was thought negligible in the conditions investigated in this study. [ABSTRACT FROM AUTHOR]

For the first time the microstructure of the oxide layer of a Zircaloy-2 cladding after 9 cycles of irradiation in a boiling water reactor has been analyzed with synchrotron micro-X-ray diffraction. Crystallographic strains of the monoclinic and to some extent of the tetragonal ZrO 2 are depicted through the thick oxide layer. Thin layers of sub-oxide at the oxide-metal interface as found for autoclave-tested samples and described in the literature, have not been observed in this material maybe resulting from irradiation damage. Shifts of selected diffraction peaks of the monoclinic oxide show that the uniform strain produced during oxidation is orientated in the lattice and displays variations along the oxide layer. Diffraction peaks and their shifts from families of diffracting planes could be translated into a virtual tensor. This virtual tensor exhibits changes through the oxide layer passing by tensile or compressive components. [ABSTRACT FROM AUTHOR]

Nanoindentation, in combination with scanning probe microscopy, has been used to measure the hardness and Young's modulus in the hydride and matrix of a high burn-up neutron-irradiated Zircaloy-2 cladding material in the temperature range 25–300 °C. The matrix hardness was found to decrease only slightly with increasing temperature while the hydride hardness was essentially constant within the temperature range. Young's modulus decreased with increasing temperature for both the hydride and the matrix of the high burn-up fuel cladding material. The hydride Young's modulus and hardness were higher than those of the matrix in the temperature range. [ABSTRACT FROM AUTHOR]

Proton-and neutron-irradiated Zircaloy-2 are compared in terms of the nano-scale chemical evolution within second phase particles (SPPs) Zr(Fe,Cr) 2 and Zr 2 (Fe,Ni). This is accomplished through ultra-high spatial resolution scanning transmission electron microscopy and the use of energy-dispersive X-ray spectroscopic methods. Fe-depletion is observed from both SPP types after irradiation with both irradiative species, but is heterogeneous in the case of Zr(Fe,Cr) 2 , predominantly from the edge region, and homogeneously in the case of Zr 2 (Fe,Ni). Further, there is evidence of a delay in the dissolution of the Zr 2 (Fe,Ni) SPP with respect to the Zr(Fe,Cr) 2 . As such, SPP dissolution results in matrix supersaturation with solute under both irradiative species and proton irradiation is considered well suited to emulate the effects of neutron irradiation in this context. The mechanisms of solute redistribution processes from SPPs and the consequences for irradiation-induced growth phenomena are discussed. [ABSTRACT FROM AUTHOR]

As one of the important structural materials in nuclear industry, the corrosion resistance of zirconium alloy limits their in-pile application. Therefore, it is necessary to investigate the corrosion mechanism of zirconium alloys. The zirconium-oxygen reaction at the O/M interface is one of the factors that affect the oxidation process. There are few reports in this regard. Ideally, the reaction process at the O/M interface has certain relevance with the initiation oxidation of zirconium, which provided a new way to investigate the reaction process by observing the initiation oxidation behaviours. To investigate the oxidation behaviours of zirconium alloy at the initial stage, in this paper, zircaloy-4 TEM thin foil specimens in 3 mm diameter were studied by TEM observation after heating in air condition with a vacuum of 3 Pa at 280 °C, 290 °C and 300 °C for 30 min exposures. The results show that, ZrO 2 begin to nucleate at a size of 3–5 nm at a high Zr/O ratio of 10.4 and oxide layer formed while Zr/O was 4.6. As a result of stress caused by the P.B ratio of Zr, slip bands formed and a bcc structure sub-oxide b-ZrOx (a = 0.51 nm) grew up along with the slip bands was observed. At both sides of b-ZrOx, two hcp structure sub-oxides having the same a -axis lattice parameter and different c -axis lattice parameter were detected. [ABSTRACT FROM AUTHOR]

Fuel swelling during normal reactor operations could lead to unfavorable chemical interactions when in contact with its cladding. As new fuel types are developed, it is crucial to understand the interaction behavior between fuel and its cladding. Diffusion experiments between U 3 Si 2 and Zricaloy-4 (Zry-4) were conducted at 800 and 1000 °C up to 100 h. The microstructure of pristine U 3 Si 2 and U 3 Si 2 /Zry-4 interdiffusion products were examined using scanning electron microscopy (SEM) and transmission electron microscopy (TEM) equipped with an energy dispersive X-ray spectroscopy (EDS) system. The primary interdiffusion product observed at 800 °C is ZrSi 2 , with secondary phases of U-Zr in the Zry-4, and Fe-Cr-W-Zr-Si phases at Zry-4/ZrSi 2 interface and Fe-Cr-U-Si phases at ZrSi 2 /U-Si interface. The primary interdiffusion products at 1000 °C were Zr 2 Si, U-Zr-Fe-Ni, U, U-Zr, and a low melting point phase U 6 Fe. [ABSTRACT FROM AUTHOR]

Dislocation structures in neutron irradiated Zircaloy-2 fuel cladding and channel material have been characterized by means of high-resolution synchrotron x-ray diffraction combined with whole peak profile analysis and by transmission electron microscopy (TEM). The samples available for this characterization were taken from high burnup fuel assemblies and offer insight into the evolution of the dislocation structure after the formation of dislocation loops containing a c component. Absolute dislocation density values are about 4–15 times higher for the whole peak profile compared to TEM analysis. Most interestingly, the diffraction analysis suggests that the total dislocation density, as well as the a loop density, increases with fluence for the cladding material type. This trend is also inferred from a Williamson-Hall representation but contradicts the TEM observations. The c loop density evolution is more complicated and doesn't display any particular trend. In addition, the diffraction analysis highlights the presence of well-developed shoulders adjacent to the basal reflections and noticeable peak asymmetry particularly for the channel samples that experienced slightly lower operation temperatures than the clad. The findings are discussed in respect of the perceived irradiation induced growth mechanisms in Zr alloys. [ABSTRACT FROM AUTHOR]

In this work, the effects of liquid subcooling, surface coating, material property, and surface oxidation on transient pool boiling heat transfer were investigated experimentally using the vertical metal rod and quenching method. The change in rod temperature was measured with time during quenching, and the visualization of boiling around the test specimen was performed using the high-speed video camera. As the test materials, the zircaloy (Zry), stainless steel (SS), niobium (Nb), and copper (Cu) were tested. In addition, the chromium-coated niobium (Cr-Nb) and chromium-coated stainless steel (Cr-SS) were prepared for accident tolerant fuel (ATF) application. Low liquid subcooling and Cr-coating shifted the quenching curve to the right, which indicates a prolongation of quenching duration. On the other hand, the material with small heat capacity and surface oxidation caused the quenching curve to move to the left. To examine the influence of the material property and surface oxidation on the film boiling heat transfer performance and minimum film boiling (MFB) point in more detail, the wall temperature and heat flux were calculated from the present transient temperature profile using the inverse heat transfer analysis, and then the curves of wall temperature and heat flux in the film boiling regime were obtained. In the present experimental conditions, the effect of material property on the film boiling heat transfer performance and MFB point seemed to be minor. On the other hand, based on the experimental results of the Cu test specimen, the surface oxidation had great impact on them. As compared with the as-received Cu test specimen, the pre-oxidized one showed the higher film boiling heat transfer performance and MFB heat flux, but it had the lower MFB temperature. [ABSTRACT FROM AUTHOR]

Radiation damage induced in Zircaloy-4 by 2.6 MeV proton irradiation at low doses was studied. Our aim is to emulate the effect of neutron irradiation during early stages of irradiation. X-ray diffraction analysis reveals that the domain size remains constant, while the microstrain increases after irradiation to a dose of 0.017 dpa. The micro-hardness, nano-hardness and Young's modulus are found to decrease after irradiation to a dose of 0.017 dpa. These indicate, respectively, the formation of a high concentration of point defects, probably vacancy type, and a decrease in the second phase precipitates size. [ABSTRACT FROM AUTHOR]

We have presented a method to prepare a uniform anodic nanoporous oxide film on the surface of a cylindrical zircaloy (Zr) tube. The distribution of the electric field around the Zr tube determines the distribution of the thickness of the anodic nanoporous oxide film. The electric field generated when a cylindrical Zr tube is electrochemically anodized was simulated by using commercial code COMSOL. When four Pt wires were used as counter electrodes, a uniform electric field was achieved with minimal use of Pt. Based on the simulation results, a cylindrical Zr tube was anodized and the distribution of the thickness of the anodic nanoporous oxide layer was measured by FESEM. Also, mass production of uniform nanoporous anodic oxide films was possible by symmetrically arranging the zircaloy tubes and Pt wires. [ABSTRACT FROM AUTHOR]

• Molten core-concrete interaction is likely to generate Fukushima Cs-particles. • Fukushima Cs-particles result from a low oxidised corium rich in Zr. • A temperature of interaction of 2000–2200 °C favours similar particles. Several radioactive releases from the containment took place during all the events leading up to Fukushima Daiichi endstate. Radionuclides released into the environment may differ in their composition or chemical form depending on the unit involved in the discharge and depending on the course of the accident. These differences may result from the path of the radionuclides from the core to the environment (e.g. through suppression pool water) and from the chemical and physical properties within the reactor core and the containment. On March 14th and 15th 2011, spherical glassy Cs-bearing microparticles (type-A CsMP) were collected for the first time. These microparticles might result from molten core-concrete interaction (MCCI) inside Unit 2, where the Zircaloy from the fuel cladding interacted with the SiO 2 of the concrete pedestal at high temperature. The result of this interaction is a significant release of silica-rich gases or aerosols from the melt pool, which condense in colder locations. Calculations using ThermoCalc with OECD/NEA's database TAF-ID version 11 were performed on the thermodynamics of vaporization during an MCCI to find correlations between the chemical and physical properties of the reactor atmosphere and the chemical composition of type-A CsMP. Data from ORIGEN2 and MELCOR calculations were used to estimate the core melt and atmosphere composition at the assumed time of the reactor pressure vessel failure. The vapour temperature at which type-A CsMP may have condensed in the containment has been evaluated. Using Fe/Si, U/Si and Cs/Si ratios in the observed CsMP as an indicator for vapour composition, this suggests that type-A CsMP originate from condensation of vapours around 2000–2200 °C from a low oxidised composition of corium, rich in zirconium and poor in stainless steel. [ABSTRACT FROM AUTHOR]

To optimize the oxidation behavior of chromium-coated Zircaloy-4 alloys under 1200 °C high-temperature steam for accident tolerant fuel (ATF) claddings, Cr coatings with different columnar structure are constructed by changing bias voltage using high power impulse magnetron sputtering and corresponding microstructure, mechanical and high-temperature steam oxidation properties are studied systematically. With the bias voltage increase from 0 V to − 150 V, the microstructure of Cr coatings changes from porous columnar structure to compact columnar structure. Especial for bias voltage is − 90 V, the tiny compact grains uniform distribute on the surface micrographs and optimal compact columnar structure are observed on the cross-section micrographs. The hardness and elastic modulus of as-prepared Cr coatings increase firstly and then decrease, the hardness of Cr coatings with optimal compact columnar structure is 8.65 GPa, which is elevated ∼2.28 times than Cr coatings with porous columnar structure. 1200 °C high-temperature steam oxidation test results show an interesting phenomenon that the thickness of Cr layer reduces initial and then increase with the exposure time prolong from 1 h to 4 h. The main steam oxidation mechanism, in the early oxidation stage, is formation the compact Cr 2 O 3 oxide layer. Then, with the exposure time extend, the reduction reaction plays the significant part for steam oxidation mechanism, which is consumption of outmost Cr 2 O 3 oxide layer and formation of subsurface Cr layer. The construction of Cr coatings with compact columnar structure and relevant oxidation mechanism are expected to offer a new idea for improving 1200 °C steam oxidation resistance of chromium-coated Zircaloy-4 alloys. • Cr coatings with different columnar structures is constructed by HiPIMS. • Hardness of optimal compact columnar structure Cr coatings is elevated ∼2.28 times. • Thickness of Cr layer becomes thicker at 1200 °C steam environment after 4 h exposure time. • The high-temperature steam oxidation mechanism of optimized Cr coatings is proposed. [ABSTRACT FROM AUTHOR]

Cr coating onto Zircaloy for the application of accident tolerance fuel claddings became a popular topic in the nuclear material field since the Fukushima-Daiichi accident in 2011. However, in the Cr coating layer the solute element effect on the potential performance improvement is still unknown, especially under irradiation. The irradiation effect in Cr-7Fe was studied after irradiation by 2.8 MeV Fe2+ to 20 dpa at 550 °C. The nano-indentation results show that doping of Fe not only can strengthen the material but also can significantly reduce the irradiation-induced hardness compared with pure Cr. The microstructure analyses show that solute element Fe can retain the diffusion of defects and reduce void swelling. In addition, the APT results indicate that no detrimental Fe-rich cluster formed in irradiated Cr-7Fe. From the aspects of irradiation-induced hardening, void-swelling, and clustering, a great improvement in irradiation resistance was achieved in Cr-7Fe alloy compared with pure Cr. [Display omitted] [ABSTRACT FROM AUTHOR]

This work reports atomic-scale experimental and modeling studies of the zirconium oxide microstructure formed at the metal-oxide (M/O) interface, and the crystallography as well as the redox chemistry of alloying element nickel in the corrosion layer found on a serviced boiling water reactor Zircaloy-2 fuel rod cladding. Complying with radioprotection aspects a small-sized sample with only limited radioactivity, taken from a very high burnup (∼ 79 MWd/kgU) cladding tube material, was prepared for the investigation using synchrotron-based X-ray microprobe techniques such as X-ray fluorescence (μXRF), X-ray diffraction (μXRD) and X-ray absorption spectroscopy (μXAS). The patterns of Ni K-edge XAS of solute Ni, probed at several location within the corrosion layer, have been recorded and analyzed. Close to the metal-oxide interface both the allotropic forms of zirconium oxide, monoclinic and tetragonal phases, have been identified by XRD analyses. XAS results of the barrier layer at the metal-oxide interface reveal that the oxides contain Ni atoms as divalent cations located in the proximity of oxygen vacancies. Near the M/O interface regions, all oxidized nickel atoms have a Ni2+ homogeneous distribution and no trace or even evidence of any metallic Ni is found in the oxide matrix. The experimental results, also supported by first-principles ligand field DFT calculations, are interpreted in term of energetics, stability, chemical and structural specificity of divalent Ni ions in zirconium oxide microstructure. The present investigation on the basis of a combined experimental and theoretical approach shows that the oxidized nickel ions being situated in the neighborhood of oxygen defects is influential on the hydrogen evolution reaction including the hydrogen intake behavior accompanied with the oxidation reaction in irradiated Zircaloy-2. • This work reports an atomic-scale study of the redox chemistry of nickel in the corrosion layer formed on a Zircaloy-2 cladding. • The chemical state and Ni speciation in the oxide layer have been obtained by synchrotron X-ray light experiments. • Near the metal/oxide interface regions, all oxidized nickel atoms have a bivalent ion distribution adjacent to oxygen vacancies. • There is no trace or even evidence of any metallic Ni found in the oxide matrix. • The results of this study have strong implications for understanding the hydrogen intake behavior in irradiated Zircaloy-2. [ABSTRACT FROM AUTHOR]

• Extended, vented dry storage of ASNF fuel is both safe and viable. • Extended, sealed dry storage of ASNF fuel is both safe and viable. • No flammability issues are identified for neither sealed nor vented canisters. • No corrosion issues are identified for neither sealed nor vented canisters. • No overpressurization issues are identified for sealed dry storage canisters. This paper presents the technical basis for extended dry storage of aluminum-clad spent nuclear fuel (ASNF). This basis addresses and closes five knowledge gaps defined by the U.S. Department of Energy (DOE) Spent Nuclear Fuel Working Group (SNFWG) in DOE-ID/RPT-1575 "Aluminum-clad Spent Nuclear Fuel: Technical Considerations and Challenges for Extended (>50 Years) Dry Storage." The SNFWG's primary concern pertaining to dry storage of ASNF is rooted in the presence of (oxy)hydroxide layers adherent to the aluminum cladding of the spent fuel that formed in service, indicating a risk of radiolytic molecular hydrogen (H 2) generation in a strong gamma radiation field—a process that can be associated with disintegration of (oxy)hydroxide and the molecular decomposition of chemisorbed and physisorbed water bonded to it. Potential consequences are overpressurization of sealed ASNF dry storage canisters and creation of a flammable atmosphere in sealed and vented canisters. Further, there is the concern that the generation of other chemical species due to chemical reactions, such as nitric acid (HNO 3), could lead to canister component corrosion. To address these issues, experimental work was conducted, including investigations of the performance of surrogate aluminum specimens and actual ASNF material featuring adherent (oxy)hydroxide layers during wet or dry storage. Further, a comprehensive characterization of the sample material was completed. Drying experiments were executed, providing better estimations of the amount of residual, chemisorbed, and physisorbed water after ASNF dry storage canister conditioning. These empirical activities allowed for the development of predictive numerical models to simulate the chemical species yield within sealed and vented dry storage canisters for 50 years. The findings of these activities lead to the main conclusion that extending current ASNF dry storage durations by 50 years is both safe and viable. [ABSTRACT FROM AUTHOR]

The formation mechanism of (01 1 ¯ 1) C14 pyramidal stacking faults (SFs) with a 8.6° deflection to the SFs on (11 1 ¯) C15 planes of Zr(Fe, Cr) 2 secondary phase particle (SPP) in Zircaloy-4 under cyclic deformation were investigated by using transmission electron microscopy. Results showed that local C14 domain structures can be induced into the C15 SPP when the SFs formed on (11 1 ¯) C15 close-packed planes by the emission of 1 6 [ 112 ] C15 Shockley partial dislocations under cyclic stress effects. Moreover, when the shear stresses along [ 11 2 ¯ ] C15 direction were large enough to initiate the 1 6 [ 11 2 ¯ ] C15 partials on every other (111) C15 planes, the C15→C14 Laves phase transformation would be accomplished, resulting in a significant reconstruction of atomic configurations of the SFs on original (11 1 ¯) C15 planes. Thus, the formation of (01 1 ¯ 1) C14 pyramidal SFs during C15→C14 phase transformation in SPP were presented. [Display omitted] [ABSTRACT FROM AUTHOR]

The zirconium (Zr) ζ-hydride phase was first proposed in 2008. While electron diffraction experiments characterized a trigonal crystal structure for this phase, formation-energy calculations have predicted that trigonal is not a stable structure for that. In this work, we attempt to tackle this discrepancy between experimental and computational studies. For this purpose, nano-beam electron diffraction (NBED) and electron energy-loss spectroscopy (EELS) were utilized to characterize the crystal structure of nano-hydrides in annealed Zircaloy-2. Diffraction patterns (DPs) collected from multiple axes of precipitates (and identical to those reported for the ζ-hydride in previous works) are shown to be composed of typical α-Zr and δ-hydride reflections along with additional reflections that did not belong to either phase. It is shown that precipitates were δ-hydride and extra reflections did not originate from a trigonal structure. Instead, it is proposed that they originated from either δ-hydride and a thin surface (probably Zr-oxide) phase or double-diffraction scattering between the α-Zr and δ-hydride. Previous works reported a plasmon energy (PE) value of 17.4 ± 0.1 eV for the ζ-phase. PE maps in this work confirmed the δ-nature of the examined nano-hydrides and revealed that the measured 17.4 eV PE value belonged to interfacial ribbons as a normal interface effect, instead of a ζ-phase. Hence, as predicted by computations, it is concluded that the trigonal structure appears to not be the appropriate crystal structure for nano-hydrides and reported observations of the ζ-hydride in the past could, in fact, have been δ-phase that was misidentified. [Display omitted] [ABSTRACT FROM AUTHOR]

The BISON fuel performance analysis code has been used to study FeCrAl and Zircaloy cladding performance to better understand cladding deformation behavior in boiling water reactors during conditions caused by anticipated operational occurrences (AOOs) or by anticipated transients without scram (ATWS). BISON models of novel separate effects tests simulating ATWS events predicted mechanical responses to thermal cycling boundary conditions. We assessed FeCrAl and Zircaloy deformation and burst models to determine those best suited for cyclic dryout event analysis. In addition, we compared diametral strain profiles of the post-experiment claddings to simulation results. The capability of the BISON code for modeling BWR AOO or ATWS events is promising. [ABSTRACT FROM AUTHOR]

• Neutron transport theory method and density functional theory method for SiC, ZrC and TiC carbides. • Neutronic, mechanical and thermodynamic behaviors under different conditions and high temperature. • Alternative nuclear fuel cladding material. In the framework of the using carbide materials as alternative nuclear fuel cladding material, an investigation has been performed to obtain neutronic, mechanical and thermodynamic behaviors under different conditions and high temperature. To achieve this goal, we have used neutron transport theory methods to study neutronic performance and density functional theory methods for mechanical and thermodynamic kinetics of carbides, SiC, ZrC and TiC. In the neutronic calculations, we have compared the results of carbides with widely used cladding materials, such as baseline Zircaloy, 304SS, 310SS, FeCrAl and APMT. On the other hand, the carbides have been analyzed from the perspective of mechanical and thermodynamic properties. The obtained lattice parameters, elastic constants, elastic modulus, Poisson's ratio, hardness, heat capacity, Debye temperature distribution, enthalpy, free energy and entropy have been discussed. The presented results advance underlying mechanisms of SiC, TiC and ZrC and guide future research, such as fuel-cladding systems, structural materials and new generation nuclear reactor systems. [ABSTRACT FROM AUTHOR]

• The coupling of pressure, thermal flux and irradiation is considered in the analysis of surface-coated Zircaloy cladding. • Two failure modes are defined when cracks are present in the coating. • A reasonable value of coating thickness should not exceed 40 μm. • The plastic behavior of material cannot be ignored in crack safety analysis of coating. In this paper, consider the coupling of pressure, thermal flux and irradiation, the mechanical response of surface-coated Zircaloy cladding with different coating materials (Cr and Fe-Cr-Al), thickness and initial crack length has been investigated. The crack length and coating thickness determine whether the crack will propagate and whether the plastic zone will reach the interface of coating and cladding. The plastic zone size and energy release rate of different coating thickness are calculated in ABAQUS. Two failure modes of coating are determined, which are 1) the plastic zone reaches the interface of cladding and coating, 2) a crack channel connecting coolant and cladding formed in coating. According to the results, the second failure mode plays a decisive role in crack safety. Different crack lengths can be divided into three zones: safety zone, discussion zone and danger zone. Safe crack lengths under different conditions are calculated. Results show that the crack tolerance of Fe-Cr-Al is better than Cr and a relatively reasonable value should not exceed 40 μm. Meanwhile, the plastic behavior cannot be ignored in calculation. Conclusions in this paper have a positive impact on the actual engineering evaluation of structural safety. They can also provide reference for future experiments. [ABSTRACT FROM AUTHOR]

• Failures of steam oxidized zircaloy clad column are studied. • Effects of various reflood rates on thermal shock assisted failure are studied. • Hydrogen generated from clad oxidation is estimated experimentally. • Maximum hydrogen shoot-up remains same irrespective of reflood rates. • Clad column has failed during reflooding in all experiments. An unmitigated loss of coolant flow accident degrades the decay heat removal from the reactor core. Because of this situation, the temperature of the clad column escalates. The steam surrounding the cladding causes an exothermic oxidation reaction between the cladding surface and steam at elevated temperatures typically above 700 °C. Under such a scenario, the fire fighting water injects into the core to reduce the accident severity and arrest the clad temperature escalation. Usually, violent heat exchanges happen between the clad column and cold water during the late reflood quenching event. Furthermore, the expansion and subsequent shrinkage of the clad column could be structurally constrained. Thus, the oxidized clad column may be exposed to unchecked thermal shock during the reflood quenching of a severely heated core, causing clad failure or fracture. In this study, a 2x2 square lattice arrangement of the zircaloy-4 cluster is selected for the experiment. This is to study the quenching behaviour of the cluster prolonged heated (1150⁰ C- 1250⁰C) clad column in steam media, simulating late phase injection of DEC-A scenario. The bottom reflooding is carried out at flow rates between 3 and 10 L per minute, typical to a light water reactor SAMG action. This study explores the qualitative structural integrity of the clad column. The analytical hydrogen generation is also estimated for all cases using the Baker and Just correlation. The experimentally measured transient hydrogen generation rate agrees well with the analytical prediction. The maximum amount of equivalent cladding reacted is 27 % during heating. A renewed enhanced oxidation leads to a drastic shoot-up in the hydrogen concentration during the reflood quenching. Nevertheless, the maximum hydrogen shoot-up remains the same irrespective of the reflood rates. The clad column in each test case has failed during reflooding. [ABSTRACT FROM AUTHOR]

CrCN/TiSiCN nano-multilayer coatings with different C/N composition ratios were prepared on Zircaloy by hybrid multi-arc ion plating and magnetron sputtering technique, and the microstructures and oxidation behaviors of the coatings in 1000 °C steam were investigated. At the C/N ratio of 1.0, the coating structure changed from a quasi-bicrystalline layer periodic arrangement to a crystalline/amorphous layer arrangement. Increasing C/N ratio inhibited the generation and mobility of N 2 bubbles in the coatings during oxidation, avoiding crack generation and coating flaking. The coating with C/N of 1.0 exhibited optimal high-temperature oxidation resistance. Moreover, the oxidation mechanisms of CrCN/TiSiCN nano-multilayer coatings were discussed. • CrCN/TiSiCN coatings with multilayer structure were deposited on Zircaloy. • The structure of as-deposited coatings changes from bicrystalline layer to crystal/amorphous layer. • Diffusion and aggregation of N 2 bubbles are suppressed by the interface of the multilayer structure. • The reasons for the difference in oxidation behavior of coatings at 1000 °C were analysed. [ABSTRACT FROM AUTHOR]

• The thermal–mechanical coupled modeling on U-10Mo HCF performance was carried out based on Abaqus/Standard finite element program. • Multi-physics behaviors were incorporated into the thermal–mechanical coupling simulation via user subroutines. • Thermomechanical responses of unirradiated and irradiated helical cruciform fuel were analyzed. • The evaluation based on U-10Mo material for HCFs have been proposed both in thermal and mechanical respects. The helical cruciform fuel (HCF) rods show an axially twisted shape with four-petaled profile, thus providing enhanced heat transfer performance and mechanical self-supporting. In this study, the fully thermal–mechanical coupled modeling was carried out to investigate the fuel performance of U-10Mo HCF rods. Abaqus subroutines were utilized to incorporate in-pile multiple physical behaviors. The thermal–mechanical responses of U-10Mo HCF were analyzed under both unirradiated and irradiated conditions. The maximum fuel temperature is estimated to be 641 K at 1.63 % FIMA burnup. The largest cladding Mises stress is estimated to be 67 MPa in the elbow zone under the unirradiated condition, and it moves to the cladding blade area and increases up to 422 MPa under the effect of irradiation swelling. Additionally, the unirradiated fuel performance under different power densities was simulated, and the mechanical contact evolution was also preliminarily analyzed. Generally, the predicted plastic strain in Zircaloy cladding brings question to feasibility of HCF compatibility with U-10Mo fuel. [ABSTRACT FROM AUTHOR]

[Display omitted] • The dominant wear mechanism of Zr-4 alloy was transformed from adhesive to abrasive wear after pre-oxidization. • ZrO 2 layer presented a transition from monoclinic to tetragonal phase after irradiation. • Fe11+ irradiation reduced ZrO 2 to Zr 2 O 3 due to the bombardment-induced Gibbs segregation (BIS) or lattice relaxation. • ZrO 2 film could promote the abrasion-resistance properties of Zr-4 alloy before and after irradiation. Metal oxide film, formed on the surface of cladding tube deriving from radiolysis of cooling water under neutron irradiation, has outstanding irradiation stability and abrasion resistance. This study tends to investigate the effect of ion irradiation on wear behaviors of pre-oxidized Zr-4 alloy. Microstrcture and mechanical properties of pre-oxidized Zr-4 alloy before and after irradiation were examined comparatively and the internal correlations between microstrcture, mechanical properties and tribological behaviors were further investigated. The results indicated that the irradiation could result in phase transition and reduction reaction of ZrO 2 layer. Raman spectra evidenced that the irradiation of pre-oxidized Zr-4 alloy with 3 MeV Fe11+ ions could result in a transformation from monoclinic ZrO 2 to tetragonal ZrO 2. XPS verified that the Fe11+ irradiation reduced ZrO 2 to Zr 2 O 3 , which could be assigned to the bombardment-induced Gibbs segregation (BIS) or lattice relaxation. Friction tests results showed that the formation of ZrO 2 film could promote the abrasion-resistance properties of Zr-4 alloy before and after irradiation owing to the increase of hardness and the change in intrinsic wear mechanism. [ABSTRACT FROM AUTHOR]

• LFA has been proved to reliably measure thermal diffusivity of Zircaloy tube. • Hydrogen plays significant roles in thermal diffusivity of Zircaloy cladding. • Hydrogen reduces the thermal diffusivity of α-Zr phase in Zircaloy. • Inflection of thermal diffusivity and electrical resistivity is due to the phase transformation. Thermal diffusivity measurements on Zirconium-based cladding materials have historically been a challenge due to the difficulty to measure on specimens with curved geometries, including nuclear grade Zircaloy cladding materials. In this work, we first used laser flash analysis method and four-probe configuration method to measure the thermal diffusivity and electrical resistivity of Zircaloy tubes respectively, which show good agreement with Zircaloy plates in this work, as well as previously published data. The consistent results proved the applicability of the laser flash analysis setup and four-probe configuration method for investigating thermal diffusivity and electrical resistivity of Zircaloy tubes. We further investigated the hydrogen effect on thermal diffusivity and electrical resistivity of Zircaloy. Hydrogen plays significant roles in the thermal diffusivity of Zircaloy, which depends on the hydrogen concentration. For higher hydrogen contents (1130 and 1820 wppm in this work), where phase transformation (α-Zr + δ-hydride → α-Zr + β-Zr) occurs at 567 °C, thermal diffusivity decreases as a function of temperature at the α-Zr + δ-hydride phase regime, while increase at higher temperature at the α-Zr + β-Zr phase regime. For low hydrogen concentration, where hydride dissolved into α-Zr matrix phase at higher temperature, the thermal diffusivity is lower than non-hydrided Zircaloy-4 with similar temperature-dependent trends. Such observation demonstrates the hydrogen effects on reducing thermal diffusivity of α-Zr phase. Hydrogen increases the electrical resistivity of Zircaloy. Similar to thermal diffusivity results, the phase transformation causes a reversal of temperature-dependent trends in electrical resistivity results in Zircaloy-4 with higher hydrogen contents. [ABSTRACT FROM AUTHOR]

Accident tolerant fuel claddings seek to improve safety margins during loss-of-coolant accident (LOCA) scenarios by reducing cladding steam oxidation rates, hydrogen production, delaying or mitigating cladding burst, and reducing cladding deformation. After rapid and extensive progress in recent years, Cr coated Zr-based alloys, ferritic FeCrAl, and SiC/SiC ceramic matrix composites (CMCs) have emerged as the most promising cladding concepts, but there are few studies comparing the accident tolerance of all three concepts simultaneously. In this study, LOCA burst testing has been conducted on Cr high impulse power magnetron sputtered Zry-4 (Cr/Zry-4), C26M (FeCrAl), and SiC/SiC SiGA (CMC). Experimental observations indicate Cr/Zry-4 burst at higher temperatures than Zry-4 however the effectiveness of the coating on mitigating ZrO 2 formation via Cr 2 O 3 formation depended on physical proximity to the burst opening. No dramatic improvement in diametrical strain nor opening geometry was observed. C26M burst at higher temperatures than Zry-4 at lower pressures, exhibiting small amounts of strain and smaller openings, and formed slow-growing Al 2 O 3 regardless of location. The CMCs did not burst or show any macroscopic signs of deformation, all the way up to an internal cladding pressure of 15.2 MPa. High temperature expanding Nb plug testing indicated that CMC room temperature strength may persist up to 1650°C, with an approximately 40% loss of strength at 1900°C. CMC steam oxidation was also compared to reference chemically vapor deposited SiC up to 1700°C, and CMCs were found to oxidize at slightly accelerated rates that were still orders of magnitude lower than bare Zry-4. Projected H 2 generation during LOCA burst testing showed that when accounting for cladding burst exposing inner diameters to steam, Cr coated Zry-4 reduced H 2 generation by a factor of two compared with bare Zry-4, while both C26M and SiC/SiC reduced H 2 generation by orders of magnitude relative to bare Zry-4. [Display omitted] [ABSTRACT FROM AUTHOR]

• The purpose of this research is to build a database composed of all simulated loss-of-coolant accident data to develop an empirical model. • Analysis of the experimental data indicates the rupture dimensions are strongly correlated to the peak balloon strain. • Develop a rupture length empirical correlation as a function of peak balloon strain and a rupture width. The dispersal of nuclear fuel through a cladding rupture opening during a transient event is the primary safety concern preventing the United States nuclear industry from extending burnup beyond a peak rod average burnup of 62 GWd/tU. The majority of research thus far has prioritized understanding, or at a minimum prioritized quantifying, fuel fragmentation and pulverization as a function of burnup and to a lesser extent onset temperature. Early high burnup loss-of-coolant accident tests performed at the Halden Boiling Water Reactor evaluated the impact of relocation on cladding temperature, whereas historical and recent cladding burst tests focused primarily on generating data related to cladding balloon and burst timing. However, very little effort has been put forth to develop a relationship for determining the rupture opening dimensions for Zircaloy cladding tubes bursting during a simulated light water reactor loss-of-coolant accident. This is significant because high burnup loss-of-coolant accident data suggest dispersal could be limited by the size of the rupture opening or rupture dimensions, thereby minimizing the safety concern. Therefore, the purpose of this research is to build a database composed of all simulated loss-of-coolant accident data to develop an empirical model that conservatively captures the rupture opening dimensions and translates those dimensions to rupture area. Analysis of the experimental data indicates the rupture dimensions are strongly correlated to the peak balloon strain. This relationship is leveraged to develop a rupture length empirical correlation as a function of peak balloon strain and a rupture width empirical correlation as a function of rupture length. Outliers in the experimental data are identified and discussed. Finally, the rupture opening empirical model is presented in such a way for general implementation into high fidelity fuel performance codes or fuel performance codes used for safety analyses. [ABSTRACT FROM AUTHOR]

The annealing behavior of the pre-oxidized Cr-coated Zry-4 at 1200 °C in argon is systematically investigated. A Cr 2 O 3 scale with a thickness of ∼9 μm formed on the sample surface after pre-oxidation of the Cr-coated Zry-4 in steam at 1200 °C for 30 min. During annealing in inert atmosphere, the thickness of the Cr 2 O 3 scale decreases with the increase of the annealing time because of the reaction between the Cr 2 O 3 scale and the outward diffused Zr at the Cr 2 O 3 /Cr interface, the decomposition of Cr 2 O 3 , and the inward diffusion of O into the Zircaloy substrate along the ZrO 2 precipitates at the Cr grain boundaries. The Cr 2 O 3 scale has completely transformed into Cr after annealing for one hour. Pores and Cr grains are observed inside the Cr 2 O 3 scale after annealing due to the decomposition of Cr 2 O 3. The ZrO 2 precipitates at Cr grain boundaries gradually grow outward during the annealing and finally reach the outer surface of the Cr coating. The results in this paper provide new insights into the thickness decrease mechanism and the decomposition behavior of the Cr 2 O 3 scale. [ABSTRACT FROM AUTHOR]

The fuel rod integrity during extended dry storage can be jeopardized by hydrogen-related phenomena such as delayed hydride cracking or embrittlement. In this work, a model for hydrogen behavior in liner claddings is proposed and validated against out-of-pile experiments. Simulations of base irradiation scenarios are carried out. The obtained results are in qualitative agreement with the post-irradiation observations. [Display omitted] [ABSTRACT FROM AUTHOR]

During operation in a nuclear reactor, Zr-based nuclear fuel cladding is subject to waterside corrosion which can lead to hydrogen ingress. The hydrogen that enters the material will migrate to colder spots and precipitate as zirconium hydrides if the hydrogen content exceeds the hydrogen terminal solid solubility in the material. Since a temperature gradient is established in the radial direction of the cladding during operation, the hydrides can preferentially precipitate at the colder outer surface of the cladding. Other gradients can also occur in the longitudinal and azimuthal directions of the cladding tube. As a consequence, hydrogen redistributes itself in response to the concentration and temperature gradients present in the sample. The response of the hydrogen in solid solution to temperature gradients is governed by the heat of transport Q* as a function of temperature, so it can be used in the BISON code. A set of experiments was set up to determine the heat of transport (Q*), in which a uniformly hydrided Zircaloy-4 sample is annealed under a fixed temperature gradient at a range of temperatures, and the resulting hydrogen distribution is analyzed to determine Q*. The results are discussed in terms of existing literature. [ABSTRACT FROM AUTHOR]

We investigated how twin modes in hexagonal close packed materials can affect large inelastic strain behavior and fracture. We considered the two twin mode systems of ( 11 2 ¯ 1 ) [ 1 ¯ 1 ¯ 26 ] and (0001) [ 1 ¯ 1 ¯ 20 ] in zircaloy-2, with each mode having 24 unique twin systems. We then incorporated these twin and parent slip systems with a dislocation-density crystalline plasticity, a non-linear finite-element, and fracture framework that accounts for crack nucleation and propagation. We investigated how these twin modes affect the interrelated effects of crack nucleation and propagation, dislocation density and inelastic slip evolution, stress accumulation, and lattice rotation. The predictions indicate that twin modes significantly affect local deformation and fracture behavior, and, therefore, are essential for the accurate representation of behavior at different physical scales in heterogeneous crystalline hexagonal close packed systems. [ABSTRACT FROM AUTHOR]

Zircaloy-4 oxide stress profiles and tetragonal:monoclinic oxide phase fraction distributions were studied using nano-beam transmission X-ray diffraction. Continuous stress relief and phase transformation during the first cycle of oxide growth was observed. The in-plane monoclinic stress was shown to relax strongly up to each transition, whereas in-plane tetragonal stress-relief (near the metal-oxide interface) was only observed post transition. The research demonstrates that plasticity in the metal and the development of a band of in-plane cracking both relax the monoclinic in-plane stress. The observations are consistent with a model of transition in which in-plane cracking becomes interlinked prior to transition. These cracks, combined with the development of cracks with a through-thickness component (driven primarily by plasticity in the metal) and/or a porous network of fine cracks (associated with phase transformation), form a percolation path through the oxide layer. The oxidising species can then percolate from the oxide surface to the metal/oxide interface, at which stage transition then ensues. [ABSTRACT FROM AUTHOR]

Samples of Zircaloy 2 have been irradiated with 4 MeV protons to two different doses. Microstructures of the unirradiated and irradiated samples have been characterized by Electron Back Scatter Diffraction (EBSD), X-ray diffraction line profile analysis (XRDLPA), Positron Annihilation Lifetime Spectroscopy (PALS) and Coincident Doppler Broadening (CDB) Spectroscopy. Tensile tests and micro hardness measurements have been carried out at room temperature to assess the changes in mechanical properties of Zircaloy 2 due to proton irradiation. The correlation of dislocation density, grain size and yield stress of the irradiated samples indicated that an increase in dislocation density due to irradiation is responsible for the change in mechanical behavior of irradiated Zircaloy. [ABSTRACT FROM AUTHOR]

Nuclear fuel designs are being developed to replace highly enriched fuel used in research and test reactors with fuels of low enrichment. In the most challenging cases, U–(7–10 wt%)Mo monolithic plate fuels are proposed. One of the considered designs includes aluminum-alloy cladding, which provides some challenges in fabrication and fuel/cladding interaction during service. Zircaloy cladding, specifically Zry–4, was investigated as an alternative cladding, and development of a fabrication method was performed by researchers with the Comisión Nacionalde Energia Atómica (CNEA) in Argentina, resulting in test fuel plates (Zry–4 clad U–7Mo) which were subsequently tested in the Advanced Test Reactor in Idaho. Because Zry–4 and U–(7–10)Mo have similar high-temperature mechanical properties, fabrication was simplified in that the fuel foil and cladding could be co-rolled and bonded. The challenge was to prevent a thermal-expansion mismatch, which could destroy the fuel/cladding bond before complete bonding was achieved; the solution was to prevent the composites from cooling significantly during or between roll passes. The final product performed very well in-reactor, showing good bonding, very little fuel/cladding interaction—either from fabrication or in-reactor testing—and little swelling, especially no detectable heterogeneous bubble formation at the fuel/cladding interface tested to a fission density of up to 2.7E+21 (average) fissions/cm 3 , 3.8E+21 (peak). [ABSTRACT FROM AUTHOR]

Results of biaxial creep tests with stress changes on Zircaloy-2 tube samples are presented. A Hollomon-type viscoplastic strain hardening model is extended by the Armstrong-Frederic nonlinear kinematic hardening law, resulting in a mixed (i.e. isotropic and kinematic ) strain hardening model. The creep tests with stress changes and similar tests published in the literature are simulated by the models. It is shown that introduction of the kinematic strain hardening in the viscoplastic strain hardening model is sufficient to describe the creep transients following stress drops, stress reversals and stress removals. [ABSTRACT FROM AUTHOR]

Fracture toughness and mechanical properties of the zircaloy-2 processed by rolling at different temperatures have been investigated, and simulations have been performed using extended finite element method (XFEM). The solutionized alloy was rolled at different temperatures for different thickness reductions (25-85%). Fracture toughness has been investigated by compact tension test. The improved fracture toughness of the rolled zircaloy-2 samples is due to high dislocation density. SEM image of the fractured surface shows the reduction in dimple sizes with the increase in dislocation density due to the formation of microvoids as a result of severe strain induced during rolling. Compact tension test, edge crack, center crack and three-point bend specimen simulations have been performed by XFEM. In XFEM, the cracks are not a part of finite element mesh and are modeled by adding enrichment function in the standard finite element displacement approximation. The XFEM results obtained for compact tension test have been found to be in good agreement with the experiment. [ABSTRACT FROM AUTHOR]

Nuclear core components of zirconium alloys experience asymmetric stress or strain cycling during service which leads to plastic strain accumulation and drastic reduction in fatigue life as well as dimensional instability of the component. Variables like loading rate, mean stress, and stress amplitude affect the influence of asymmetric loading. In the present investigation asymmetric stress controlled fatigue tests were conducted with mean stress from 80 to 150 MPa, stress amplitude from 270 to 340 MPa and stress rate from 30 to 750 MPa/s to study the process of plastic strain accumulation and its effect on fatigue life of Zircaloy-2 at room temperature. It was observed that with increase in mean stress and stress amplitude accumulation of ratcheting strain was increased and fatigue life was reduced. However, increase in stress rate led to improvement in fatigue life due to less accumulation of ratcheting strain. [ABSTRACT FROM AUTHOR]

The most direct means of improving the ability of water cooled reactors to withstand excessive cladding oxidation during a loss of coolant accident is to either modify or replace zirconium cladding. It is important to understand what level of agreement is to be expected as a function of systematic differences in steam oxidation testing techniques and instrumentation among testing facilities. The present study was designed to assess the sensitivities of some of the current and proposed reactor cladding materials. Steam oxidation sensitivity of Zircaloy-2, FeCrAl and Mo to O 2 impurities in steam were examined. It was shown that the effect of O 2 impurities is negligible for the two former materials while significant in the case of Mo. [ABSTRACT FROM AUTHOR]

The effect of rolling at different temperatures and deformation strains on the variation in hardness and tensile properties of Zircaloy-4 has been investigated in the present work. The initial hot extruded and annealed Zircaloy-4 was subjected to various rolling reductions of 25%, 50%, 70%, and 90% at room (298 K) and liquid nitrogen (77 K) temperatures along extrusion and transverse directions of the as-received bar, respectively. The hardness and ultimate tensile strength (UTS) obtained for samples rolled to a true strain of 2.3 along the extrusion direction were 264.67 ± 3.8 HV, 786 MPa at 298 K, and 273.57 ± 1.6 HV, 856 MPa at 77 K, while along the transverse direction, the hardness and UTS values observed were 262.7 ± 3.77 HV, 727 MPa at 298 K, and 266.57 ± 2.7 HV, 827.63 MPa at 77 K. The enhancement in strength and hardness of the Zr alloy is attributed to increase in the stacking fault probability (SFP) with the applied progressive true strains at different temperatures, which led to the greater accumulation of dislocation density and hence a better grain refinement. [ABSTRACT FROM AUTHOR]

Electrochemical corrosion of Zircaloy-2 was tested at room temperature in lithium hydroxide (LiOH) concentrations that ranged from 2.2 to 7000 ppm and boric acid (H 3 BO 3 ) concentrations that ranged from 50 to 4000 ppm. Following the corrosion experiments, the oxide films of specimens were examined by SEM to examine the oxide existence. LiOH concentrations as high as 1 M (7000-ppm lithium) can lead to significantly increased electrochemical corrosion rate. It is suggested that the accelerated corrosion in concentrated solution is caused by the synergetic effect of LiOH, pH and porosity generation. In solutions containing 100 ppm of lithium, the presence of boron had an ameliorating effect on the corrosion rates of Zircaloy-2. Similar to acceleration of corrosion by lithium, the inhibition by boron is due to a combined effect of pH neutralizing and its participation in the corrosion process. [ABSTRACT FROM AUTHOR]

In the present work, the mechanical behavior of Zircaloy-4 subjected to various deformation strains by multiaxial forging (MAF) at cryogenic temperature (CT) was investigated. The alloy was strained up to different number of cycles, viz., 6 cycles, 9 cycles, and 12 cycles at cumulative strains of 2.96, 4.44, and 5.91, respectively. The mechanical properties of the alloy were investigated by performing the universal tensile test and the Vickers hardness test. Both the test showed improvement in the ultimate tensile strength and hardness value by 51% and 26%, respectively, at the highest cumulative strain of 5.91. The electron backscattered diffraction (EBSD) measurement and transmission electron microscopy (TEM) were used for analyzing the deformed microstructure. The microstructures of the alloy underwent deformation at various cumulative strains/cycles showed grain refinement with the evolution of shear and twin bands that were highest for the alloy deformed at the highest number of cycles. The effective grain refinement was due to twins formation and their intersection, which led to the improvement in mechanical properties of the MAFed alloy, as observed in the present work. [ABSTRACT FROM AUTHOR]

TEM thin foil specimens prepared using Zircaloy-4 coarse-grained samples were corroded by autoclaving at 250 °C and 280 °C in lithated water for a short time exposure, and the initial oxidation process before the formation of ZrO 2 was investigated by TEM. The results show a variation of the crystal structure along with the increase of oxygen contents at the initial oxidation stage. When the Zr/O atomic ratio reached 6–7, 3 and 1, a commensurable long period super-lattice 9a 0 -2H and sub-oxides with hcp and fcc ordered structure were formed. When the Zr/O atomic ratio was 0.8, monoclinic ZrO 2 was detected. [ABSTRACT FROM AUTHOR]

Pd-Ni/Al2O3 systems were investigated in the reaction of hydrogen oxidation in terms of their possible application as catalysts used in passive autocatalytic recombiners (PARs) used in nuclear power plants. Testing experiments were carried out in a flowing system at different temperatures and humidity of the reaction mixture. The bimetallic catalysts exhibited higher response to the increase of temperature and higher resistance to inhibit water than the monometallic palladium catalyst. They showed excellent stability during a few tens of hours, similarly, like their monometallic counterpart. Our bimetallic catalysts of hydrogen oxidation can be used as cheaper alternatives to catalysts based on the precious metals in the hydrogen oxidation without loss of their activity over time. [ABSTRACT FROM AUTHOR]

Iron-chromium-aluminum (FeCrAl) coatings deposited on Zircaloy 2 (Zy2) and yttria-stabilized zirconia (YSZ) by magnetron sputtering have been tested with respect to oxidation weight gain in high-temperature steam. In addition, autoclave testing of FeCrAl-coated Zy2 coupons under pressure-temperature-dissolved oxygen coolant conditions representative of a boiling water reactor (BWR) environment has been performed. Four different FeCrAl compositions have been tested in 700 °C steam; compositions that promote alumina formation inhibited oxidation of the underlying Zy2. Parabolic growth kinetics of alumina on FeCrAl-coated Zy2 is quantified via elemental depth profiling. Autoclave testing under normal BWR operating conditions (288 °C, 9.5 MPa with normal water chemistry) up to 20 days demonstrates observable weight gain over uncoated Zy2 simultaneously exposed to the same environment. However, no FeCrAl film degradation was observed. The 900 °C eutectic in binary Fe–Zr is addressed with the FeCrAl-YSZ system. [ABSTRACT FROM AUTHOR]

Abstracts Burst behavior of austenitic and ferritic Fe-based alloy tubes has been examined under a simulated large break loss of coolant accident. Specifically, type 304 stainless steel (304SS) and oxidation resistant FeCrAl tubes were studied alongside Zircaloy-2 and Zircaloy-4 that are considered reference fuel cladding materials. Following the burst test, characterization of the cladding materials was carried out to gain insights regarding the integral burst behavior. Given the widespread availability of a comprehensive set of thermo-mechanical data at elevated temperatures for 304SS, a modeling framework was implemented to simulate the various processes that affect burst behavior in this Fe-based alloy. The most important conclusion is that cladding ballooning due to creep is negligible for Fe-based alloys. Thus, unlike Zr-based alloys, cladding cross-sectional area remains largely unchanged up to the point of burst. Therefore, for a given rod internal pressure, the temperature onset of burst in Fe-based alloys appears to be simply a function of the alloy's ultimate tensile strength, particularly at high rod internal pressures. [ABSTRACT FROM AUTHOR]

The chemistry of barrier oxide layers formed on zirconium alloys was investigated using atom probe tomography (APT). Grain boundary segregation of Fe and Ni was observed. The Zr(Fe, Cr) 2 particles maintain a constant Zr:Fe:Cr ratio through the oxide until they reach an oxygen content of approximately 50 at%, when Fe depletion occurs. Enrichment of hydrogen along oxide crystallite boundaries was observed, which is interpreted as a sign of ingress being localized to grain boundaries. TEM revealed porosity networks around particles and oxide grain boundaries. It is proposed that pores are local reduction sites where H 2 evolves on transition metal sites. [ABSTRACT FROM AUTHOR]

The article discusses the importance of maintaining safety in the nuclear power plants and the severe damages of the boiling water reactors in the Fukushima-Daiichi nuclear power plants due to earthquakes and the tsunamis after it. It also talks about the use of Zircaloy-2 as the cladding materials for the reactors. It discusses the accident tolerant fuel for forms to be used in the light water reactors.